• Nuclear Engineering and Technology
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• v.22 no.1
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• pp.12-19
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• 1990

The Loss of Offsite Power Transient at 77.5% power which occurred on June 9, 1981 at the Kori Unit 1 PWR (Pressurized Water Reactor) is simulated using the RELAP5/MOD2 system thermal-hydraulics computer code. Major thermal-hydraulic parameters are compared with the available plant data. The comparison of the analysis results with the plant data demonstrates that the RELAP5/MOD2 code has the capability to simulate the thermal-hydraulic behaviour of PWRs under accident conditions of this type with accuracy, except the pressurizer pressure and level. The pressurizer pressure increase is sensitive to the in surge now it is believed that the interracial heat transfer in a horizontal stratified flow regime may be estimated low and the compression effect due to insurge flow may be high. In the nodalization sensitivity study it is found that S/G noding with junctions between bypass plenum and steam dome is preferred to simulate the S/G water level decreasing and avoid the spurious level peak at trubine trip.

• Nuclear Engineering and Technology
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• v.22 no.1
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• pp.66-74
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• 1990

The thermal-hydraulic characteristics of the 17$\times$17 OFA (Optimized Fuel Assembly) used in the KNU 7&8 are analyzed and compared with that of the 17$\times$17 SFA (Standard Fuel Assembly) loaded in the KNU 5&6. The thermal-hydraulic characteristics analyzed are minimum DNBR, fuel centerline temperature and exit void fraction at normal operation and design over power transient. Additionally, local linear rod power, which will cause fuel centerline melting, is calculated. The DNBR sensitivity calculations are performed with respect to the reactor operating parameters. COBRA-IV-I code is used for these calculations. The modified W-3 correltion and the drift-flux model are applied for the critical heat flux calculation and the void fraction calculation, respectively. From the calculated results, it has been found that the possibility of DNB occurrence is higher in the OFA than in the SFA. The other hand, the local linear power resulting in fuel centerline moiling of the OFA is nearly equal to that of the SFA.

• Nuclear Engineering and Technology
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• v.56 no.10
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• pp.4412-4422
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• 2024

The CFETR multipurpose overload robot (CMOR) is a critical component of the fusion reactor remote handling system. To accurately calculate and visualize the structural deformation and stress characteristics of the CMOR motion process, this paper first establishes a CMOR kinematic model to analyze the unfolding and working process in the vacuum chamber. Then, the dynamic model of CMOR is established using the Lagrangian method, and the rigid-flexible coupling modeling of CMOR links and joints is achieved using the finite element method and the linear spring damping equivalent model. The co-simulation results of the CMOR rigid-flexible coupled model show that when the end load is 2000 kg, the extreme value of the end-effector position error is more than 0.12 m, and the maximum stress value is 1.85 × 10⁸ Pa. To utilize the stress-strain data of CMOR, this paper designs a CMOR morphology prediction control system based on Unity software. Implanting CMOR finite element analysis data into the Unity environment, researchers can monitor the stress strain generated by different motion trajectories of the CMOR robotic arm in the control system. It provides a platform for subsequent research on CMOR error compensation and extreme operation warnings.

• Transactions of the Korean Society of Mechanical Engineers B
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• v.24 no.11
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• pp.1447-1456
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• 2000

Simulated experiments of high pressure melt ejection(HPME) are performed to measure the released fraction of corium simulant from the French type PWR cavity. The experiments are carried out on a 1/20th linear scaled model of the Ulchin 1&2 cavity. Water or woods metal and nitrogen is used as simulant of molten corium and steam, respectively. Experimental parameters are water mass, annulus area and breach size. It is shown that only breach size effects is very important while the mass and the annulus area do not affect the released fraction. It is found that the liquid film transport is much more dominant mechanism than the entrainment droplet transport, especially in linear scale down simulated HPME experiment.

• Journal of Energy Engineering
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• v.22 no.3
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• pp.283-286
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• 2013

The existing nuclear power plants have been often redesigned by increasing or decreasing electrical power without changing design concept by the request of utility, economic factors or other factors. When the cost of power plant equipment redesigned by changing reactor power and electrical power is estimated, if its quotation is not available in the market place, cost scaling factor(CSF) applies to the cost of existing plant equipment and then the new-designed equipment cost can be calculated. In this paper, we review CSFs according to plant capacity change cases in United State DOE, EPRI, ABB, SWEC and introduce the results applied to Korean PWR 1000MWe and 1400MWe.

• Journal of Information Processing Systems
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• v.1 no.1 s.1
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• pp.36-40
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• 2005

TRISO (Tri-Isotropic)-coated fuel particle is widely applied due to its higher stability at high temperature and its efficient retention capability for fission products in the HTGR (high temperature gas-cooled reactor), one of the highly efficient Generation IV reactors. The typical ball-type TRISO-coated fuel particle with a diameter of about 1 mm is composed of a nuclear fuel particle as a kernel and of outer coating layers. The coating layers consist of a buffer PyC, inner PyC, SiC, and outer PyC layer. In this study, a digital image processing algorithm is proposed to automatically measure the thickness of the coating layers. An FBP (filtered backprojection) algorithm was applied to reconstruct the CT image using virtual X-ray radiographic images for a simulated TRISO-coated fuel particle. The automatic measurement algorithm was developed to measure the coating thickness for the reconstructed image with noises. The boundary lines were automatically detected, then the coating thickness was circularly by the algorithm. The simulation result showed that the measurement error rate was less than 1.4%.

• Proceedings of the Korea Society for Energy Engineering kosee Conference
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• 2000.11a
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• pp.115-120
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• 2000

The analysis of flux distribution under stead-state in large power reactors with assymetry reactivity insertions requires the use of three-dimensional diffusion calculations. For the purpose, consistently formulated modern nodal methods based on higher order interface techniques have become popular tools for flux distributions in large commercial nuclear reactors. Among the earlier developments, the nodal Green's function method obtains its nodal interface equation from the transverse-integrated integral diffusion equation using a finite-medium Green's function. In this method, the outgoing current from a node surface is formulated as a response of the incoming currents and the spatially integrated neutron source within the same node. The well-known nodal expansion method is also based on an interface partial current formulation. Nodal methods high-level interface variables, i.e., interface net current and flux, may be more computationally efficient than the nodal Green's function method because they have one fewer unknown per interface. The Analytic Nodal Method(ANM), which can be classified as an interface net current technique and, was faster in solving some standard benchmark problems than the other two methods.(omitted)

• Journal of KIISE:Software and Applications
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• v.32 no.12
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• pp.1178-1191
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• 2005

In this paper, we propose a synthesis method of software fault tree from software requirements specification written in NuSCR formal specification language. The software fault tree, proposed in this paper, reflects requirements on both structure and behavior and it is an integrated form. The software fault tree can be used for analyzing safety in the view of structure and behavior. We propose templates for each components in NuSCR specification language and a synthesis method of software fault tree using the templates. The research was applied into the main trip logic of the reactor protection system of ARP1400, the Korean next generation nuclear reactor system, developed by KNICS. And we evaluate feasibility of our approach through this case study.

• Proceedings of the ESK Conference
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• 1997.04a
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• pp.56-60
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• 1997

The study of human erroneous actions has traditionally taken place along two different lines of approach. One has been concerned with finding and explaining the causes of erroneous actions, such as studies in the psychology of "error". The other has been concerned with the qualitative and quantitative prediction of possible erroneous actions, exemplified by the field of human reliability analysis (HRA). Another distinction is also that the former approach has been dominated by an academic point of view, hence emphasising theories, models, and experiments, while the latter has been of a more pragmatic nature, hence putting greater emphasis on data and methods. We have been developing a method to make predictions about error modes. The input to the method is a detailed task description of a set of scenarios for an experiment. This description is then analysed to characterise thd nature of the individual task steps, as well as the conditions under which they must be carried out. The task steps are expressed in terms of a predefined set of cognitive activity types. Following that each task step is examined in terms of a systematic classification of possible error modes and the likely error modes are identified. This effectively constitutes a qualitative analysis of the possibilities for erroneous action in a given task. In order to evaluate the accuracy of the predictions, the data from a large scale experiment were analysed. The experiment used the full-scale nuclear power plant simulator in the Halden Man-Machine Systems Laboratory (HAMMLAB) and used six crews of systematic performance observations by experts using a pre-defined task description, as well as audio and video recordings. The purpose of the analysis was to determine how well the predictions matiched the actually observed performance failures. The results indicated a very acceptable rate of accuracy. The emphasis in this experiment has been to develop a practical method for qualitative performance prediction, i.e., a method that did not require too many resources or specialised human factors knowledge. If such methods are to become practical tools, it is important that they are valid, reliable, and robust.

• Journal of the Korea Institute of Information and Communication Engineering
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• v.19 no.1
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• pp.185-191
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• 2015

This paper suggest the safety class communication board in order to design the safety network of the nuclear safety class controller. The reactor protection system use the digitized networks because from analog system to digital system. The communication board shall be provided to pass the required performance and test of the safety class in the digital network used in the nuclear safety class. Communication protocol is composed of physical layer(PHY), data link layer(MAC: Medium Access Control), the application layer in the OSI 7 layer only. The data link layer data package for the cyber security has changed. CRC32 were used for data quality and the using one way communication, not requests and not responses for receiving data, does not affect the nuclear safety system. It has been designed in accordance with requirements, design, verification and procedure for the approving the nuclear safety class. For hardware verification such as electromagnetic test, aging test, inspection, burn-in test, seismic test and environmental test in was performed. FPGA firmware to verify compliance with the life-cycle of IEEE 1074 was performed by the component testing and integration testing.